1. Technical Field
The present invention relates to a zirconium alloy and a method of preparing the same. More particularly, the present invention relates to a method of preparing a zirconium alloy having a low hydrogen pick-up rate and high hydrogen embrittlement resistance, and to a zirconium alloy composition having a low hydrogen pick-up rate and high hydrogen embrittlement resistance.
2. Description of the Related Art
Zirconium alloys are used for nuclear fuel cladding tubes, guide tubes and spacer grids as nuclear fuel component in a nuclear power plant. Under the operation environments of a nuclear power plant, the mechanical properties of zirconium alloys deteriorate because of high-temperature and high-pressure corrosion environment and neutron irradiation. Zirconium, as a raw material of zirconium alloys, has a very small neutron absorption cross section, excellent high-temperature strength and corrosion resistance, and is widely used in a nuclear reactor core in the form of an alloy containing a small amount of niobium, iron, chromium or the like.
Among conventional zirconium alloys, zircaloy-2 and zircaloy-4 containing tin, iron, chromium and nickel are most widely used. Currently, ZIRLO, which a zirconium alloy prepared by adding a small amount of niobium, iron, chromium, etc. to zirconium, is used world-wide.
However, recently, as part of the improvement in economic efficiency of nuclear reactor, a high burn-up for extending nuclear fuel cycles has been used within a severe environment, and thus the reaction time of nuclear fuel with high-temperature and high-pressure cooling water increases, thereby causing problems of nuclear fuel corrosion and hydrogen embrittlement. Owing to a hydrogen pick-up by zirconium alloy corrosion, the hydrides in zirconium matrix are formed and the strength of the zirconium alloy becomes very low due to delayed hydride cracking (DHC) and fracture toughness deterioration by hydride.
Therefore, it is necessary to develop a zirconium alloy having excellent corrosion resistance and hydrogen embrittlement resistance under primary cooling water environment with a high-temperature and high-pressure in a nuclear power plant. For this reason, research into developing a zirconium alloy having high corrosion resistance and a low hydrogen pick-up rate has been variously conducted. In this case, since the optimization conditions for providing a low hydrogen pick-up rate and high hydrogen embrittlement resistance to a zirconium alloy are influenced by the kind and amount of added elements, processing condition, heat treatment condition or the like, it is most important to establish an alloy design and an alloy preparation process.
The paper “Zirconium Alloy E635 as a Material for Fuel rod Cladding and Other Components of VVER and RBMK Cores, 11th International Symposium on Zirconium in the Nuclear Industry, ASTM STP 1295, eds. by Bradley and Sabol, pp. 785˜803, written by Nikulina et al.” discloses the fact that a zirconium alloy, prepared by adding 0.95˜1.05 wt % of niobium, 1.2˜1.3 wt % of tin and 0.34˜0.4 wt % of iron to zirconium, has very excellent corrosion resistance when the ingot of the zirconium alloy is β-annealed at 900˜1070° C., water-cooled, α-pressed at 600˜650° C., cold-worked and intermediate-heat-treated (heat treatment temperature: 560˜620° C.) three to four times, and then final-treated at 560˜620° C.
U.S. Pat. No. 4,938,920 discloses a zirconium alloy composition including niobium 0˜1.0 wt %, tin 0˜0.8 wt %, vanadium 0˜0.3 wt %, iron 0.2˜0.8 wt %, chromium 0˜0.4 wt %, oxygen 0.1˜0.16 wt % and a residue of zirconium. Here, when the total amount of chromium and vanadium is adjusted within a range of 0.25˜1.0 wt %, this zirconium alloy has higher corrosion resistance than that of zircaloy-4.
U.S. Pat. No. 5,254,308 discloses a zirconium alloy composition having improved corrosion resistance and a hydrogen pick-up rate, including niobium 0.015˜0.3 wt %, tin 1.0˜2.0 wt %, iron 0.07˜0.7 wt %, chromium 0.05˜0.15 wt %, nickel 0.16˜0.4 wt %, silicon 0.002˜0.050 wt %, oxygen 0.09˜0.16 wt % and a residue of zirconium. Here, the zirconium alloy composition is configured such that the ratio of iron and chromium is 1.5, the amount of added niobium is determined depending on the additive amount of iron influencing a hydrogen pick-up rate, and the additive amount of nickel, silicon, carbon or oxygen is determined to allow the zirconium alloy to have excellent corrosion resistance and strength.
U.S. Pat. No. 5,648,995 discloses a method of fabricating a nuclear fuel cladding tube using a zirconium alloy including niobium 0.8˜1.3 wt %, iron 50˜250 ppm, oxygen 1600 ppm or less, silicon 120 ppm or less. In this method, a nuclear fuel cladding tube was fabricated by the steps of: primarily heat-treating the zirconium alloy at 1000˜1200° C., β-quenching the zirconium alloy, secondarily heat-treating the quenched zirconium alloy and then extruding this zirconium alloy; cold-rolling the extruded zirconium alloy 4˜5 times; intermediate-heat-treating the cold-rolled zirconium alloy at a temperature range of 565˜605° C. for 2˜4 hours between the cold-rolling steps; and final-heat-treating this zirconium alloy at 580° C. In this case, in order to improve the creep resistance of the nuclear fuel cladding tube, the content of iron in the zirconium alloy is restricted within a range of 250 ppm or less, and the content of oxygen in the zirconium alloy is restricted within a range of 1000˜1600 ppm.
U.S. Pat. No. 5,940,464 discloses a process of preparing a zirconium alloy including niobium 0.9˜1.1 wt %, tin 0.25˜0.35 wt %, iron 0.2˜0.3 wt %, carbon 30˜180 ppm, silicon 10˜120 ppm, oxygen 600˜1800 ppm and a residue of zirconium. The prepared zirconium alloy was heat-treated at 1000˜1200° C., quenched, drawn at 600˜800° C., and then heat-treated at 590˜650° C. Subsequently, this zirconium alloy was cold-rolled four or more times, and then intermediate-heat-treated at 560˜620° C. between the cold rolling steps. Thereafter, this zirconium alloy was finally cold-rolled, and then finally heat-treated by recrystallization annealing (RXA, 560˜620° C.) and stress relief annealing (SRA, 470˜500° C.).
As described above, research for improving the corrosion resistance, hydrogen embrittlement resistance and a low hydrogen pick-up rate of zirconium alloys used for nuclear reactor core materials such as nuclear fuel cladding tubes and the like has been variously conducted. It is continuously required to develop zirconium alloys having a low hydrogen pick-up rate and high hydrogen embrittlement resistance that can maintain the stability of nuclear fuel during a high burn-up efficiency long-period operation.
Therefore, as the present inventors have conducted research for developing new zirconium alloys as replacements for conventional zirconium alloys, they have found that new zirconium alloy compositions have lower hydrogen pick-up rate and higher hydrogen embrittlement resistance than those of conventional zirconium alloy compositions. Based on this finding, the present invention has been created.